Real-time reflectometry – An ASDEX Upgrade DCS plugin App for plasma position and shape feedback control

A multi-loop plasma position and current control system in the Globus-M tokamak (Ioffe Institute, St. Petersburg, Russia) has been developed and modelled on a new evolution plasma code designed in MATLAB. Robust PID controllers in the plasma position control loops were tuned by the Quantitative Feedback Theory (QFT) based on the identified linear model of a nonlinear single-phase thyristor current inverter as a self-oscillating actuator.

PPCC is hybrid group of interacting controllers managed through a decision-making logic. After nine years of successfully controlling the plasma in all JET experiments, it is being overhauled to integrate the new full boundary controller Extreme Shape Controller. The difficult integration issues have spawn a re-evaluation of the overall management strategy of the many control abilities provided by PPCC. The aim of this paper is to describe the complex real-time decision logic implemented in PPCC and to explain how it has been shaped by the needs of JET plant. Furthermore, it will focus on the changes necessary to integrate the new component.

The renewed HT-7 plasma control system based on real-time Linux cluster

It is pointed out that plasma position and shape control calls for fast algorithms for plasma boundary identification, i.e. methods capable of determining the shape in a time interval not longer than the sampling time of the controller. The authors present an online identification algorithm which realizes a trade-off between accurate plasma shape reconstruction and computational burden. The proposed identification has been tested with reference to some types of ITER (International Thermonuclear Experimental Reactor) plasmas.< <ETX xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">&gt;</ETX>

Plasma control at JET

マルチプロセッサシステムを用いた核融合プラズマ位置形状制御システムの開発

This paper describes a multi-winding transformer as a plasma initiator in tokamak devices and the application to the design of poloidal field coil power supply system. The scale of tokamak device has been considerably increased based on some experimentally obtained scalings of a energy confinement, because the plasma performance has been improved with the current and size parameters. Since large tokamaks require the hybrid functional poloidal field coil system, the each coil power supplies must prepare the high voltage generation circuit for the plasma breakdown, and it causes increase in the number of current interrupters, disconnectors and the resisters. Therefore, the plasma position and shape control tend to be difficult especially in the start-up phase of the discharge. The plasma initiator proposed in this paper makes the plasma initiation circuit to be simple by unification of the high voltage generation circuit. It has similar ampere-turn and permeance distribution with the tokamak device and produces the magnetic field necessary for plasma breakdown. An application to FER tokamak proposed by JAERI is also presented.

Currently, the fast evolution of cameras in recent years has made them promising tools for diagnostics of Tokamak. The solution presented in this paper consists of a prototype of high-speed visible image acquisition and processing system (HVIAPs) dedicated for experimental advanced superconducting tokamak shape and position control. Graphics processing unit (GPU) and field-programmable gate array (FPGA) are typically used as accelerators or co-processors in addition to a CPU. Such a heterogeneous computing system can combine the advantages of its individual components. The main imaging equipment components used for this system are four high-resolution fast cameras equipped with fiber interface. Image data from the cameras are received by the frame grabber card based on the FPGA and transmitted to the GPU via the peripheral component interconnect express interface. The software support for the system includes low-level drivers and application programming interface libraries for all components and the algorithm developed for image processing. The offline image process results are compared to equilibrium fitting code, which is a commonly used reconstruction method, with an average error of 1.5 cm. The total processing time for one frame is less than 0.3 ms.

Model-based plasma vertical stabilization and position control at EAST

This paper deals with the problem of designing a plasma current, position, and shape controller for the ITER-FEAT tokamak. Exploiting the features of the plant, we show that it is possible to design separately two control loops: a first loop which stabilizes the vertical position by means of a simple derivative action, and a second loop which drives the plasma current and the geometrical shape descriptors as close as possible to the reference values. The second loop is designed by assigning the dominant closed loop poles.

A laser produced plasma light source for a small field exposure tool (SFET) has been developed at the EUVA Hiratsuka R&D center. The light source consists of the following components: The drive laser of the xenon plasma source is a short-pulse, high-power KrF laser that has been developed in cooperation with Gigaphoton Inc. and Komatsu Ltd. The laser has an unstable resonator and produces a maximum output power of 580W at 4kHz repetition rate. The xenon target is a 50 micrometer diameter liquid jet with a speed of about 35 m/s. The source has been designed to generate 0.5W in-band power at the intermediate focus (IF) at a collecting solid angle of pi sr. The source includes automatic control, e.g. jet and plasma position control, and an electrical interface for the exposure tool. The performance of the source at IF has been evaluated by Canon Inc. This paper explains source performances. Especially, results of IF parameters like image size, position stability and out of band radiation are presented.

A quasi-steady-state alternating current operation assisted by lower hybrid wave (LHW) was achieved on a HT-7 superconducting tokamak with plasma current of Ip = 125 kA, line-averaged density of 1.5 × 1019 m−3, electron temperature of Te = 500 eV and 30–50 s plasma duration. Plasma current was sustained and smoothly transferred from one direction to the other without loss of ionization. Plasma position control, LHW assistance, strong gas puffing and good wall condition are the key issues to have a smooth transition of plasma current. Our modelling results show that current reversal equilibrium configuration with two oppositely flowing currents in the high-field-side and the low-field-side during current reversal exists. This is in agreement with experimental measurements.

Toroidal effects on the outer magnetic surfaces of a configuration having a large toroidal plasma current that is stabilized by a pitch reversal produced by an external helical coil (OHTE) are investigated by both field-line tracing and an analytic expansion of the magnetic surfaces. Results of the two methods are compared. It is shown that both plasma position control and the helical coil winding law can be used to compensate undesired toroidal effects at both low and high beta.

An algorithm is presented that identifies major confinement regimes on-line during a plasma discharge in the Tokamak ASDEX Upgrade. Apart from the ohmically heated regime with no additional heating power, low and high radiative L and H regimes are distinguished. The algorithm was successfully implemented in the ASDEX Upgrade discharge control system, allowing a dynamic control of the plasma discharge. In a first application, the control system recognized the falling back of the plasma from a high confinement regime to a low confinement regime and reacted accordingly, avoiding a disruption and enhancing the plasma performance by recovering the high confinement regime.

Nuclear fusion is, in a sense, the opposite of nuclear fission. Fission, which is a mature technology, produces energy through the splitting of heavy atoms like uranium in controlled chain reactions. Unfortunately, the by-products of fission are highly radioactive and long lasting. On the other hand, fusion is the process by which the nuclei of two light atoms such as hydrogen are fused together to form a heavier (helium) nucleus, with energy produced as a by-product. Although controlled fusion is extremely technologically challenging, a fusion-power reactor would offer significant advantages over existing energy sources. This thesis is devoted to the control of tokamaks, magnetic confinement devices constructed in the shape of a torus (or doughnut). Tokamaks are the most promising of several proposed magnetic confinement devices. The need to improve the performance of modern tokamak operations has led to a further development of the plasma shape and position control systems. In particular, extremely elongated plasmas, with high vertical-instability growth rate, are envisaged to reach the required performance for ignition. This request for better performance from the experimentalists’ side has motivated the development of the new vertical-stabilization (VS) system at the JET tokamak, which has been proposed within the Plasma Control Upgrade project. This thesis presents the activity carried out to increase the capability of the VS system and to understand the operational limits in order to assess what can be done to improve the overall performance with the existing hardware and control system so as to minimize the impact on JET operation. The first objective of this work is the analysis of the new diagnostic system and the influence of the mechanical structure on the magnetic measurements used as diagnostics by the VS controller; the main focus is on the influence on the controller performance in the presence of large perturbations. The second objective is to design a new controlled variable to increase the performance of the VS system. The third objective is to provide an equivalent model of an ELM (Edge Localized Mode), in terms of internal plasma profile parameters via best fit of the vertical velocity estimation. The last objective is to obtain a reliable and accurate model of the overall system, based on the new platform MARTe, developed at JET and useful also for other devices.